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2024 31st International Conference on Nuclear Engineering, ICONE 2024

作者： Zhu, Ye Xingbo, Wang Xianwei, Liao Zhiyun, Cai Minghao, Liu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9780791888247

The reactor coolant system (RCS) is the primary system of the nuclear power plant primary cycle. Some small module reactors (SMR) applied once-through steam generator and integrated reactor vessel. The characteristic of these reactors differs widely from the tradition nuclear power plant. APROS Nuclear simulation and analysis software, with the Comprehensive plant model covering reactor island, turbine island, electrical and automation systems, has been successfully used in a series of major nuclear power plant projects, including power upgrade, modernization, safety improvement, and new plant projects. A reactor coolant system model of SMR that contains both process and control elements is developed and simulated using APROS. Steady states analysis is conducted under different power load and both force/natural cycle. Analysis of Transient condition including increasing/decreasing power load and force/natural circulation switching is also conducted to acquire dynamic thermal hydrate parameter. Reference and guidance can be provided for system optimization and safety analysis. Copyright © 2024 by ASME.

关键词： Nuclear power plants

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RESEARCH ON THE OPTIMIZATION OF THE reactor COOLANT system STRUCTURE 30

RESEARCH ON THE OPTIMIZATION OF THE REACTOR COOLANT SYSTEM S...

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30th International Congress on Sound and Vibration, ICSV 2024

作者： Yanli, Yuan YeXianhui Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Sichuan Chengdu China

ISBN: (纸本)9789090390581

The dynamic analysis of reactor coolant system is the key step of mechanics evaluation, and with the continuous improvement of seismic design benchmarks and system design requirements, it is urgent to carry out the research on seismic capacity optimization of reactor coolant system. The reactor coolant system has the character that the design parameters are numerous and the structure is nonlinear, and when the structure is optimized, it is difficult to determine the key parameters which affect the dynamic response of the reactor coolant system prominently, which is also the difficulty in the traditional optimization according to the designer’s experience. In this paper, the sensitivity analysis of reactor coolant system dynamic response is carried out based on the OPTIMUS integrated platform first, and the key design parameters affecting the dynamic response of reactor coolant system are identified;then the optimization model for dynamic response of reactor coolant system is established according to the sensitivity analysis results, taking the support stiffness and support gap as the design variables,taking the nozzle loads of the main equipment as the objective function;finally, the sequential quadratic programming method is adopted to solve the optimization problem. The optimization analysis results show that the nozzle loads of the main equipment can be effectively decreased when appropriate design parameters are chosen. Through the research, a set of optimization analysis methods and processes applicable to the reactor coolant system and generalizable to other systems and equipment in the reactor were established,which can provide data support for the improvement of reactor system. © 2024 Proceedings of the International Congress on Sound and Vibration. All rights reserved.

关键词： Sensitivity analysis

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Research on Hybrid Modeling Method Based on Mechanism Model and BP Neural Network 4

Research on Hybrid Modeling Method Based on Mechanism Model ...

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4th Asia Conference on Information Engineering, ACIE 2024

作者： Chen, Jie Zheng, Yanqiu Chu, Qing Jiang, Guanfu Li, Jing Nuclear Power Institute of China Science and Technology on Reactor System Design Technology Laboratory Chengdu610213 China

ISBN: (纸本)9798350371925

reactor system modeling is the premise of reactor control system design. The common commercial thermal and hydraulic analysis programs like RELAP5, RETRAN have complex structure and slow operational speed, which are not conductive to the transplantation and improvement of reactor system. In order to realize the rapid development of reactor control system, this paper carries out the mechanism modeling of the reactor system in Matlab/Simulink, and then introduce BP neural network to establish the hybrid model, which improves the modeling accuracy to support the design of reactor system. For small pressurized water reactor, we firstly take analytical modeling method to build the core mechanism model and its state space model through three conservation principles and the operating mechanism of core. Based on analyses, we can conclude that the parameter error and structural error of the mechanism model is caused by the assumption and simplification during modeling, and choose the parameter with the most influence (the feedback coefficient of the coolant temperature) as the corrected parameter in the BP neural-based hybrid modeling process. After that, according to the dual constant operation scheme of small pressurized water reactors, and the gradient descent method of BP network, the hybrid model of core is built and the parameter error and structure error of the mechanism modeling are corrected online. The simulation results show that the proposed hybrid modeling method can accurately simulate the reactor operate state. Thus, the proposed hybrid model is accurate and effective. © 2024 IEEE.

关键词： Pressurized water reactors

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Transient fuel performance analysis of UO_(2)–BeO fuel with composite SiC coated with Cr cladding based on multiphysics method

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Nuclear science and Techniques 2023年第12期34卷 93-108页

作者： Chun‑Yu Yin Shi‑Xin Gao Sheng‑Yu Liu Rong Liu Guang‑Hui Su Li‑Bo Qian School of Nuclear Science and Technology Xi’an Jiaotong UniversityXi’an 710049China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China School of Electric Power South China University of TechnologyGuangzhou 510640China

The transient multiphysics models were updated in CAMPUS to evaluate the accident-tolerant fuel performance under accident *** is a fuel performance code developed based on *** simulated results of the UO_(2)–Zircaloy fuel performance under accident conditions were compared with those of the FRAPTRAN code and the experimental data to verify the correctness of the updated ***,multiphysics models of the UO_(2)–BeO fuel and composite SiC coated with Cr(SiC_(f)/SiC-Cr)cladding were implemented in ***,the fuel performance of the three types of fuel cladding systems under Loss of Coolant Accident(LOCA)and Reactivity Insertion Accident(RIA)conditions was evaluated and compared,including the temperature distribution,stress distribution,pressure evolution,and cladding failure *** results showed that the fuel temperature of the UO_(2) fuel under accident conditions without pre-irradiation was lower after being combined with SiC_(f)/SiC-Cr ***,the centerline and outer surface temperatures of the UO_(2)–BeO fuel combined with SiC_(f)/SiC-Cr cladding reduced further under accident *** cladding temperature increased after the combination with the SiC_(f)/SiC-Cr cladding under accident conditions with *** addition,the use of SiC_(f)/SiC-Cr cladding significantly reduced the cladding hoop strain and plenum pressure.

关键词： Accident condition Fuel performance UO2–BeO fuel SiCf/SiC-Cr cladding

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A Tutorial on Quantized Feedback Control

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IEEE/CAA Journal of Automatica Sinica 2024年第1期11卷 5-17页

作者： Minyue Fu IEEE the School of System Design and Intelligent Manufacturing Shenzhen Key Laboratory of Control Theory and Intelligent Systems and Centre for Control Science and TechnologySouthern University of Science and TechnologyShenzhen 518055China

In this tutorial paper, we explore the field of quantized feedback control, which has gained significant attention due to the growing prevalence of networked control systems. These systems require the transmission of feedback information, such as measurements and control signals, over digital networks, presenting novel challenges in estimation and control design. Our examination encompasses various topics, including the minimal information needed for effective feedback control, the design of quantizers, strategies for quantized control design and estimation,achieving consensus control with quantized data, and the pursuit of high-precision tracking using quantized measurements.

关键词： Consensus control high-precision control networked control quantized estimation quantized feedback control robust control

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Fretting Wear Characteristics of Nuclear Fuel Cladding in High-Temperature Pressurized Water

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Chinese Journal of Mechanical Engineering 2023年第4期36卷 326-338页

作者： Jun Wang Haojie Li Zhengyang Li Yujie Lei Quanyao Ren Yongjun Jiao Zhenbing Cai Key Lab of Advanced Technologies of Materials Tribology Research InstituteSouthwest Jiaotong UniversityChengdu 610031China Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu 610213China

In pressurized water reactor(PWR),fretting wear is one of the main causes of fuel assembly ***,the operation condition of cladding is complex and harsh.A unique fretting damage test equipment was developed and tested to simulate the fretting damage evolution process of cladding in the PWR *** can simulate the fretting wear experiment of PWR under different temperatures(maximum temperature is 350℃),displacement amplitude,vibration frequency,and normal *** fretting wear behavior of Zr-4 alloy under different temperature environments was *** addition,the evolution of wear scar morphology,profile,and wear volume was studied using an optical microscope(OM),scanning electron microscopy(SEM),and a 3D white light *** show that higher water temperature evidently decreased the cladding wear volume,the wear mechanism of Zr-4 cladding changed from abrasive wear to adhesive wear and the formation of an oxide layer on the wear scar reduced the wear volume and maximum wear depth.

关键词： Fretting wear Cladding High temperature and high pressure Zirconium alloy

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High-Order Fully Actuated system Models for Strict-Feedback systems With Increasing Dimensions

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IEEE/CAA Journal of Automatica Sinica 2024年第12期11卷 2451-2462页

作者： Xiang Xu Guang-Ren Duan Shenzhen Key Laboratory of Control Theory and Intelligent Systems Southern University of Science and TechnologyShenzhen 518055 the School of System Design and Intelligent Manufacturing Southern University of Science and TechnologyShenzhen 518055China

This paper mainly addresses control problems of strict-feedback systems(SFSs)with increasing *** with the commonly-considered SFSs where the subsystems have the same dimension,we aim to handle more complex cases,i.e.,the subsystems in the considered SFSs are assumed to have increasing *** transforming the systems into highorder fully-actuated system(HOFAS)models,the stabilizing controllers can be directly *** first-order SFSs,secondorder and high-order SFSs are also considered.

关键词： Increasing dimensions high-order fully-actuated system nonlinear systems strict-feedback systems

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design and Simulation Verification of Intelligent Perception and Adaptive Control system for Control Rod Drive Mechanism 30

Design and Simulation Verification of Intelligent Perception...

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30th International Conference on Nuclear Engineering, ICONE 2023

作者： Yang, Hui Lin, Yuanfeng Zeng, Hui Huang, Qingyu Zhang, Siyuan Yang, Kunlin Wang, Yuanmei Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9784888982566

Nuclear power plant status monitoring and equipment fault diagnosis are important guarantees for the safe operation of nuclear power plants. With the development of artificial intelligence technology, fault diagnosis of nuclear reactor equipment has entered the stage of intelligent diagnosis and prediction. The control rod drive mechanism (CRDM) is the actuator of the reactor control and nuclear safety protection system. In the normal state, CRDM realizes the stepping movement of the control rod through the alternating action of the three sets of coil currents. When an abnormality in the coil current is detected, it means that the CRDM movement fails, and in more serious cases, it may cause the control rod to accidentally drop. Therefore, in order to prevent the occurrence of control rod stuck, sliding rod and driving failure in the reactor and to ensure the safe and economic operation of the reactor, a necessary measure is to effectively monitor the operation status of the CRDM. In this paper, the overall architecture of an intelligent perception and adaptive control system for CRDM is studied. By analyzing the multi-condition and multi-mode coupling relationship in the time dimension and spatial correlation between the coil current and vibration signals, a condition discrimination and fault diagnosis model for CRDM based on hybrid network of CNN+BiGRU is established. Furthermore, a digital adaptive control model of control rod drive system is built based on the real-time simulation system dSPACE. Through functional verification, the system can perform real-time perception and identification of normal operation conditions, fault diagnosis and adaptive control based on the coil current and vibration signals. The recognition accuracy rate for various operation conditions reaches more than 99.35%, and the control accuracy of the sequential current is improved around ±4% of the rated value under various external environment changes, providing a supportive reference for

关键词： Failure analysis

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Atomic Scale Simulation on Liquid Metal Embrittlement Induced by Segregation of Lead Element in Lead-Cooled Fast reactor 23rd

Atomic Scale Simulation on Liquid Metal Embrittlement Induce...

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23rd Pacific Basin Nuclear Conference, PBNC 2022

作者： Bida, Zhu Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789811988981

Lead-Cooled fast reactor, which uses liquid lead or lead-bismuth alloy as coolants, is one of the six main reactors of the fourth generation reactor system. Most recently, the Lead-Cooled fast reactor received a lot of attention due to its safety, economy, and sustainability advantages. However, the liquid metal embrittlement (LME) could be a potential threat to the reliability of structure materials used in the Lead-Cooled fast reactor. Several experimental results show that the segregation of alloy element in grain boundary has a close association with liquid metal embrittlement. In this paper, by applying atomic simulation techniques, including molecular dynamics and Monte Carlo method, the segregation of lead element on poly-crystal ferritic and the resulting mechanical properties variations are simulated with the newest interatomic potential. The simulation results show that the lead element segregates on grain boundary. Moreover, such segregation of lead on grain boundary induced hardening and embrittlement of poly-crystal systems in the following uniaxial tensile testes. The atomic-scale mechanisms for such mechanical properties variation induced by lead element segregation are discussed carefully. In conclusion, these results verify that the segregation of lead element could be one of the critical mechanisms for LME, which well agree with previous experimental evidences. The present study may allow us to understand the mechanism of LME better. It can serve as a good reference for selecting more suitable materials for the Lead-Cooled fast reactor. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Grain boundaries

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design and Implementation of DCS design List Automatic Generation and Check system Based on LabVIEW 7th

Design and Implementation of DCS Design List Automatic Gener...

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7th International Symposium on Digital Instrumentation and Control technology for Nuclear Power Plant, SICPNPP 2023

作者： Deng, Xiao-Jun Zhang, Xu Feng, Shi-Man Peng, Hao Yao, Ying-Fan Science and Technology on Reactor System Design Technology Laboratory Nuclear Power Institute of China Chengdu China

ISBN: (纸本)9789819934546

The DCS configuration design process involves a large number of variables and multiple lists, and designers need to spend more time on the preparation, checking and maintenance of the lists. Manual compilation of files is a lot of work, and prone to human error. In this process, the use of automatic software to assist in the design can effectively improve efficiency. Based on the practical requirements of safety-class DCS design project of a nuclear power plant unit 1 and 2, this paper designs and implements a set of automatic generating and checking system of accounting list, which implements the automatic generation and typesetting of list files, and obtains good application effect. © 2023, The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd.

关键词： Nuclear power plants

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